1. Field of the Invention
The present invention relates generally to treatment of high alkaline radioactive and hazardous wastes, and more particularly to processes for immobilizing a waste containing one or more of radionuclides, hazardous elements, hazardous compounds, and other compounds present in the waste.
2. Background of the Invention
The use of radioactive and hazardous materials in the world has led to the accumulation of a significant amount of radioactive and hazardous wastes. There is an international consensus regarding the planned disposal of these wastes by burying them in the ground in deep geological repositories. At the present time, high-level radioactive wastes are being placed in long-term storage awaiting permanent disposal. Once buried, with the passage of time, groundwater and hydrothermal solutions can make contact with the radionuclides, hazardous elements, or hazardous compounds contained in the wastes. As a result, groundwater and hydrothermal solutions can facilitate the leaching of radionuclides, hazardous elements, and hazardous compounds out of the wastes into the biosphere in which plants and animals live. In addition, even without the interference from groundwater and hydrothermal solutions, radionuclides, hazardous elements, or hazardous compounds could possibly diffuse out of the wastes, resulting in contamination of the biosphere. Therefore, improper containment of the wastes can create a significant problem.
There are a number of existing processes that can potentially reduce the leaching and/or diffusion of radioactive and hazardous wastes. The existing processes, however, have various disadvantages. For example, cementation is commonly used to immobilize low-level and intermediate-level radioactive waste. This process is undesirable because a large volume of cement is required to immobilize a small quantity of wastes vastly increasing the size of the disposal area. Furthermore, cement is highly susceptible to both leaching and diffusion.
The most common method of handling high-level radioactive wastes is vitrification in borosilicate glass. Vitrification is currently being used in a number of countries including France, the United States of America, Korea, Italy, Germany, the United Kingdom, Japan, Belgium, China, and Russia. Conventional vitrification processes, however, are limited in the amount of waste that can be contained, and efforts to increase waste loading capacity of borosilicate glasses or melts have led to high crystallinity, increased rates of leaching, and increased corrosion of the melter, rendering the compositions unsuitable for use in conventional vitrification melters.
Thus, a need exists for improved vitrification processes and borosilicate glass-like compositions that achieve higher waste loading without the above-mentioned disadvantages on use of the processes and compositions with conventional vitrification melters.
Although low-level radioactive waste is not generally vitrified, in the U.S., for example, low-level radioactive waste with high sodium content is planned to be vitrified at the Waste Treatment Plant in Richland, Washington State. Sodium is a high active alkaline element, which, when immobilized in borosilicate glass in high concentrations, causes lack of durability of the glass. Much of the low-level radioactive waste found in the U.S. radioactive waste sites, such as Envelope A of the Hanford LAW (the majority of the LAW to be processed at the Hanford Waste Treatment Plant) is characterized by high sodium concentrations. The currently planned sodium waste loading for these low-level radioactive wastes is between about 20-23 weight percent, although the actual acceptable waste as reported in industry is between about 18-20 weight percent. When greater amounts of sodium are introduced into the final glass, it becomes unstable and is unable to satisfy the waste form acceptance criteria and the processing requirements for a vitrification melter.
In order to increase the concentration of sodium in the final glass, one must create a composition that will incorporate this sodium yet satisfy the waste form acceptance criteria and the processing requirements for a vitrification melter. The waste form acceptance criteria (durability and leach resistance of the glass) are tested by Vapor Hydration Test (VHT), Product Consistency Test (PCT) and Toxicity Characteristic Leach Procedure (TCLP). The processing requirements include viscosity, specific electrical conductivity and crystallization.
Thus, a need exists for improved vitrification processes and borosilicate glass-like compositions that achieve higher sodium waste loading without the above-mentioned disadvantages on use of the processes and compositions with conventional vitrification melters.